This application relates to stainless steel alloys, and in particular to stainless steel alloys having a high resistance to corrosion and stress corrosion cracking in high-temperature water. As used herein, the term "high-temperature water" means water of about 150.degree. C. or greater, steam, or the condensate thereof. As used herein, the term "stress corrosion cracking" means cracking propagated by static or dynamic stressing in combination with corrosion at the crack tip.
High-temperature water can be found in a variety of known apparatus, such as water deaerators, nuclear reactors, and in steam driven central station power generation. Corrosion and stress corrosion cracking are known phenomena occurring in the components, including structural members, piping, fasteners, and weld deposits, of apparatus exposed to high-temperature water. For example, the components in nuclear reactors exposed to high-temperature water are known to undergo stress corrosion cracking. The reactor components are subject to a variety of stresses associated with, e.g., differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources including residual stress from welding, cold work and other asymmetric metal treatments. In addition, water chemistry, welding, heat treatment, and radiation can increase the susceptibility of a component to stress corrosion cracking of the metal.
Irradiation of stainless steel alloys in the core of nuclear reactors can promote stress corrosion cracking from the segregation of impurities, such as phosphorus, silicon and sulfur, to the grain boundaries. Irradiation-assisted stress corrosion cracking has been reduced by restricting such impurities in stainless steel alloys. Thus, modified forms of such alloys as 348, 316, and 304 stainless steel (using the official classification system of the American Society of Testing and Materials) have been developed with upper limits on phosphorus, silicon and sulfur below the limits of the standard alloys. In U.S. Pat. No. 4,836,976, further reduction in susceptibility to irradiation-assisted stress corrosion cracking was achieved by limiting the nitrogen content of austenitic stainless steels to a maximum of 0.05 weight percent.
Corrosion that leads to stress corrosion cracking has been widely studied and a number of papers have been written concerning it. Some of the publications addressing stress corrosion cracking and incorporated by reference herein are:
1) F. P. Ford, "Stress Corrosion Cracking", in Corrosion Processes, edited by R. N. Parkins, Applied Science Publishers, New York, 1982, p. 271.
2) J. N. Kass and R. L. Cowan, "Hydrogen Water Chemistry TechnoIogy for BWRs", in Proc. 2nd Int. Conf on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Monterey, Calif., 1985, p. 211.
3) M. E Indig, B. M. Gordon, R. B. Davis and J. E. Weber, "Evaluation of In-Reactor IntergranuIar Stress" in Proc. 2nd Int. Conf on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Monterey, Calif., 1985, p. 411.
b 4) F. P. Ford, P. L. Andresen, H. D. Solomon, G. M. Gordon, S. Ranganath, D. Weinstein, and R. Pathania, "Application of Water Chemistry Control, On-Line Monitoring and Crack Growth Rate Models for Improved BWR Materials Performance", Proc. Fourth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Jekyll Island, Ga., August 1989, Nace, Houston, pp 4-26 to 4-51, 1990.
5) L. W. Niedrach and W. H. Stoddard, "Corrosion Potentials and Corrosion Behavior of AISI304 Stainless Steel In High Temperature Water Containing Both Dissolved Hydrogen and Oxygen", Corrosion, Vol. 42, No. 12 (1986) page 696.
It is well documented that stress corrosion cracking occurs at higher rates when oxygen is present in the reactor water in concentrations of about 5 parts per billion, ppb, or greater. Stress corrosion cracking is further increased in a high radiation flux where oxidizing species, such as oxygen, hydrogen peroxide, and short-lived radicals are produced from radiolytic decomposition of the reactor water. Such oxidizing species increase the electrochemical corrosion potential of metals. Electrochemical corrosion is caused by a flow of electrons from anodic and cathodic areas on metallic surfaces. The corrosion potential is a measure of the thermodynamic tendency for corrosion phenomena to occur, and is a fundamental parameter in determining rates of, e.g., stress corrosion cracking, corrosion fatigue, corrosion film thickening, and general corrosion.
As explained in these and other articles, stress corrosion cracking in boiling water nuclear reactors and the associated water circulation piping has been reduced by injecting hydrogen in the water circulated therein. The injected hydrogen reduces oxidizing species in the water, such as dissolved oxygen, and as a result lowers the corrosion potential of metals in the water. However, factors such as variations in water flow rates and the time or intensity of exposure to neutron or gamma radiation result in the production of oxidizing species at different levels in different reactors. Thus, varying amounts of hydrogen have been required to reduce the level of oxidizing species sufficiently to maintain the corrosion potential below a critical potential required for protection from the stress corrosion cracking in the high-temperature water.
As used herein, the term, "critical potential" means a corrosion potential at or below a range of values of about -230 to -300 mV based on the standard hydrogen electrode (she) scale. Below the critical potential, stress corrosion cracking is markedly reduced or even eliminated as disclosed in references 2-5. Stress corrosion cracking proceeds at an accelerated rate in systems in which the electrochemical potential is above the critical potential, and at a substantially lower rate in systems in which the electrochemical potential is below the critical potential. Water containing oxidizing species such as oxygen increases the corrosion potential of metals exposed to the water above the critical potential, while water with little or no oxidizing species present results in corrosion potentials below the critical potential.
Corrosion potentials of stainless steels in contact with reactor water containing oxidizing species can be reduced below the critical potential by injection of hydrogen into the water in a concentration of about 50 to 100 ppb or greater. Much higher hydrogen injection levels are necessary to reduce the corrosion potential within the high radiation flux of the reactor core, or when oxidizing cationic impurities, e.g., cupric ion are present. Such hydrogen injection lowers the concentration of dissolved oxidizing species in the water and also the corrosion potential of the metal. However, high hydrogen additions, for example of about 150 ppb or greater, that reduce the corrosion potential below the critical potential can result in a higher radiation level in the steam driven turbine section from incorporation of the short-live N.sup.16 species. The higher radiation requires additional shielding, and radiation exposure control.
Accordingly, although the addition of hydrogen lowers the corrosion potential of reactor water, it is also desirable to limit the amount of hydrogen in reactor water, while maintaining the corrosion potential below the critical potential.
In the article, "Increasing Passivation Ability and Corrosion Resistance of Stainless Steel by Surface Alloying with Palladium," G. P. Chernova, T. A. Fedosceva, L. P. Kornienko, and N. D. Tomashov, Prot. Met. (Eng. Transl.) 17 (1981) page 406, the electrochemical behavior and increase in corrosion potential and passivation of stainless steel that is surface alloyed with palladium, and exposed to deaerated acidic solutions is disclosed. The increased corrosion potential causes a passive oxide layer to form on the stainless steel that reduces further corrosion.
One object of this invention is to provide a stainless steel alloy having improved resistance to corrosion and stress corrosion cracking in high-temperature water.
Another object is to provide a stainless steel alloy comprised of high-chromium that reduces corrosion of grain boundaries within components formed from the alloy and exposed to high-temperature water.
Another object is to provide a high-chromium stainless steel alloy comprised of titanium, tantalum, niobium, or mixtures thereof that reduces corrosion of grain boundaries within components formed from the alloy and exposed to high-temperature water.
Another object is to provide a high-chromium stainless steel alloy comprised of a platinum group metal that reduces the corrosion potential of the alloy in high-temperature water.
Another object is to provide a method for reducing stress corrosion cracking of a component exposed to high-temperature water by lowering the corrosion potential of the component.